j!t Final Report

8/11/2019 j!t Final Report

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A

REPORT ON VOCATIONAL TRAINING

NUCLEAR POWER CORPORATION OF INDIA LTD.

(A Government of India Enterprise)

Rajasthan Atomic Power Station, Rawatbhata

SUBMITTED BY: JITENDRA MEENA

B.TECHVII sem (Electronics & Communication)

University Roll No.- 11EAXECO35

Submitted in the partial ful fi llment for award the degree ofBachelor of Technology

Submitted to

Department of Electronic & Communication Engineering

Apex Institute of Engineering & TechnologySession 2014 2015
http://c/Users/J!T/Downloads/LAN_RAPS34/rapp34.htmhttp://c/Users/J!T/Downloads/LAN_RAPS34/rapp34.htm


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A

REPORT ON VOCATIONAL TRAINING

FROM

NUCLEAR POWER CORPORATION OF INDIA LTD.

(A Government of India Enterprise)


SUBM ITTED TO: SUBM ITTED BY:

MR.AKHILESH SINGH JITENDRA MEENA

CORDINATOR (11EAXEC035)

SESSION 2014-15


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TABLE OF CONTENTS

Certificate from the Institute..ii

Certificate from Company/organization.....iii

Declaration from student...iv

Acknowledgement.v

Company Detail...vi

Assessment of student.vii

Performance report of student...x

Preface..xi

Table of contents xii

Conclusion..xiv

Reference.........xv


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PREFACE

As we know that an engineer has to serve an industry, for that one must be aware of industrial

environment, their management, problems and the way of working out their solutions at the

industry.

After the completion of the course an engineer must have knowledge of interrelation between

the theory and the practical. For this, one must be familiar with the practical knowledge with

theory aspects.

To aware with practical knowledge the engineering courses provides a six weeks industrial

training where we get the opportunity to get theory applying for running the various process

and production in the industry.

I have been lucky enough to get a chance for undergoing this training at RAJASTHAN

ATOMIC POWER STATION. It is a constituent of board of NPCIL. This report has been

prepared on the basis of knowledge acquired by me during my training period of 45 days at

the plant.

JITENDRA MEENA


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ACKNOWLEDGEMENT

It was highly educative and interactive to take training at

RAJASTHANATOMIC POWER STATION. As technical knowledge is incomplete

without practical knowledge, I couldnt find any place better than this to update myself.

I am very much thankful to the Site director Mr. C.P. Jhamb&Training superintendent

Mr. D. Chanda for allowing me for the industrial training at RAPS. Thanks to Mr. A.P.

Jain for their guidance during my project.

I also take the opportunity to thanks Nuclear training Centre for providing lecture on

overview of the plant and providing me Orange qualification.

JITENDRA MEENA


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INTRODUCTION

India's Nuclear power developments are under the purview of the Nuclear Power

Corporation of India, a government-owned entity under the Department of Atomic

Energy India. The corporation is responsible for designing, constructing, and

operating nuclear-power plants. In 1995 there were nine operational plants with a

potential total capacity of 1,800 megawatts, about 3 percent of India's total power

generation. There are two units each in Tarapur, north of Bombay in Maharashtra; in

Rawatbhata in Rajasthan; in Kalpakkam near Madras in Tamil Nadu; and in Narora in

Uttar Pradesh; and one unit in Kakrapur in southeastern Gujarat. However, of the nine

plants, all have been faced with safety problems that have shut down reactors for

periods ranging from months to years. The Rajasthan Atomic Power Station in

Rawatbhata, India was closed indefinitely, as of February 1995. Moreover,

environmental problems, caused by radiation leaks, have cropped up in communities

near Rawatbhata. Other plants operate at only a fraction of their capacity, and some

foreign experts consider them the most inefficient nuclear-power plants in the world.


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MISSION

To develop nuclear power technology andproduce in a self-reliant manner nuclear

power as a safe, environmentally benign

and an economically viable source of

electrical energy to meet the growing

electricity needs of the country

**********


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VISION

NPCIL has its vision to have aninstalled nuclear power capacity

of 20,000 MW(e) by the year

2020. This capacity could be

achieved by the development of

more 220 MW(e) & 550 MW(e)

units of Pressurized heavy water

reactors, importing light waterreactors and by the introduction

of fast breeder reactors.

**************


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Prime Minister

DAEAtomic Energy Commission

NPCIL

Atomic energy Regulatory

board

TAPS 1&2

India Rare Earth

TAPP 3&4

BARC

RAPS 1& 2

Heavy Water Board

RAPS 3 & 4

ECIL

RAPP 5 & 6

UCIL

MAPS

Nuclear fuel com lex

NAPS

Indra Gandhi center foradvance research

KAPS

KAIGA PS 1& 2

Center for advance technology

KAIGA Proj. 3& 4

KKPS


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RAPS LOCATION AND SITE CONDITIONS

RAPS is located on the eastern bank of Rana Pratap Sagar lake (R.P.S) upstream of

the R.P.S dam across the chambal river at an elevation of 388 mt. above mean sea

level with a latitude of 24053 north and a longitude of 76036 east. The plant site is

about 64 KM from Kota city. The place has an average rainfall of 825mm as per

records. The maximum wind velocity records so far is 129 km/hr at 120 m. the most

predominant wind direction is at 7.90m and 120m heights is North of south west and

west of south west respectively.

The site has no population with in its vicinity of radius of 5km. It however does have

a population of about 58 thousand distributed in the radius of 15 Km. the only nearby

major industry is HEAVY WATER PLANT (H.W.P).

NUCLEAR ENERGY:

Mass defect converted into energy through nuclear reaction. Two processes produce

this:

1) Nuclear fission.

2) Nuclear fusion.

A neutron it splits into two big parts hits when a heavy nucleus likes that of uranium

235 & in addition 2 or 3 neutrons are released. However, the mass of the parts is

slightly less than the mass of the uranium nucleus. The mass that is destroyed is

converted into energy (200Mev/ fission). This process is called nuclear fission

reaction.

It is much more likely if neutrons are slow, in a reactor, some of the neutrons

produced are absorbed so that for every neutron causing fission, only one is left. This

neutron in turn collides with another U235 nucleus & causes fission. A chain reaction

is thus set up. Also, the neutrons have to be slowed down. The fuel in a nuclear

reactor consists of Uranium that may be natural or enriched in which proportion of


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U235 is increased. Either light water (for enriched uranium) or heavy water (for

natural uranium) may be used as a moderator, for slowing down the neutrons. The

energy released is absorbed by the water (either light or heavy). This coolant in turn

transfers its energy to the light water. Ultimately water is turned into steam at high

pressure that is used to derive turbines as in any conventional power plant. India has

six Nuclear Power Plants;

At Tarapur in Maharastra.

At Rawatbhata near Kota in Rajasthan

At Kalpakkam near Madras in Tamil Nadu.

At Narora in Uttar Pradesh

At Kakarpara near Surat in Gujarat

At Kaiga near Karwar in Karnataka.

The reactors at Tarapur use enriched uranium as fuel & light water as moderator and

coolant, all others use uranium and heavy water. Nuclear Power Plant under

construction is two units of 500 MW at Tarapur and two similar units at Rawatbhata

near Kota. Nuclear fission has become commercially viable and is being exploited in

several countries.


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SOME IMPORTANT NUCLEAR REACTIONS:

1) 92U238+0n

1-----92U239+r------93Np

239-------94Pu239

Typical fission reaction:

2) 92U235+0n

1------38Sr90+54Xe

144+20n1+r+200MeV

Reactor poisoning reaction:

3) 52Te135----53I

135-----54Xe135-----55Cs

135---56Ba135

(Stable)

We know that about 200MeV of energy is released during per fission.

This energy is divided in the following way:

1) K.E. of the fission fragments: 167MeV.

2) K.E. of neutrons: 5MeV.

3) Energy of gamma released at fission: 5MeV.

4) Energy of gamma rays released on ncapture: 10MeV.

5) Gamma decay energy: 7MeV.

6) Betadecay energy: 5MeV.

----------------------------

TOTAL =199MeV.

----------------------------


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THREE STAGES OF INDIAN NUCLEAR POWER

PROGRAMME:

1) INTRODUCTION:

India figured in the nuclear power map of the world in 1969, when two boiling

water reactors (BWRS) were commissioned at Tarapur (TAPS-1&2) these

reactors were built on the turnkey basis .The main objective of setting these units

was, largely to prove the techno-economic viability of nuclear power.

The nuclear power programme formulated embarked on the

three-stage nuclear power programme, linking the fuel cycle of pressurized heavy

water reactor (PHWR) & Fast breeder reactors (FBR) for judicious utilization of

our reserves of Uranium & Thorium. The emphasis of the programme is self

reliance and thorium utilization as a long -term objective.

The three stages of our Nuclear power programme are:

1) STAGE I:- This stage envisages construction of natural Uranium, Heavy

water moderator & cooled pressurized heavy water reactors (PHWR). Spent fuel

from these reactors is reprocessed to obtain plutonium.

2) STAGE II: - This stage envisages on the construction of Fast breeder reactors(FBR) fuelled by plutonium & depleted U produced in stage I. These reactors

would also breed U233 from thorium.

3) STAGE III = This stage would comprise power reactors using U233-

Thorium as fuel, which is used as a blanket in these type of reactors.


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The PHWR was chosen due to the following:

1) It uses natural uranium as fuel. Use of natural uranium available in India,

helps cut heavy investments on enrichments, as uranium enrichment is capital

intensive.

2) Uranium requirement is the lowest & plutonium production is the highest.

3) The infrastructure available in the country is suitable for undertaking

manufacture of the equipment.

The short term goal of the programme was to complement the generation of

electricity at locations away from coalmines. The long-term policy is based on

recycling nuclear fuel and harnessing the available Thorium resources to meet

countrys long-term energy demand and security.

As a part of PHWR Programme (STAGE I) second nuclear power plant was

taken up as a joint Indo-Canadian venture this plant was built at Rawatbhata

(Rajasthan) two units laid a milestone in the history of India all the components

were taken up in India and the import content reduced considerably. Moreover,

Canadians withdrew in 1974; Indian engineers did balance design &

commissioning of the Unit 2.


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2) CHALLENGES FACED:

The industry was new to the manufacturing techniques & stringent quality

requirements of the nuclear components like calandria, end shield, steam

generators, fuelling machine, and heavy water pumps. The requirement of

convectional power plant equipment was of much larger capacity than those being

manufactured in the country.

To achieve self sufficiency in this field in the long run, the

department of atomic energy established extensive research & development

facilities covering diverse areas for supporting technology absorption. Facilities,

from prospecting to mining to fabrication of fuel & zirconium alloy components,

for manufacture of precision reactor components & production of heavy water

were also set up. Supply of equipments of international nuclear standard was also

a problem so momentous efforts were put into development of such

manufacturing industries. Extensive R&D set up were established for

metallurgical studies of both fresh as well as radioactive material, non

destructive testing, environmental & seismic qualification of safety analysis,

preparation &development of validation of computer codes, etc.

Technologies for inspection of the reactor components, repair &replacement using

robotics & life extension programme of the operating reactors, have also been

successfully developed.

To summaries, the concerted efforts put in by DAE, its constituent units &

NPCIL, together with Indian industries & institutions have led to development &full capabilities to design, manufacturing of equipment, construction, operation &

maintenance of nuclear power plant.

Today India is amongst the select band of few countries of the world that

have developed such capabilities.

3. Status of nuclear power generation & future plans:


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The nuclear power programme in India up to year 2020 is based on

installation of a series of 235 MWe &500Mwe pressurized heavy water reactor

(PHWR) UNITS, 1000MWe light water reactors (LWR) UNITS & fast breeder

reactors (FBR) units. NPCIL plans to contribute about 10% of the total additional

needs of power of about 10000MWe per year i.e. 1000 MWe per year in the

coming two five year plans. The total installed capacity of nuclear generation

would increase to more than 20000 MWe in year 2020 from the present level of

2720 MWe.

The basic design of the 220/500MWe units in similar; however, a

number of significant design changes have been made progressively from the first

unit at Rajasthan to the 500 MWe units. These design changes have been made

from the consideration of currently prevailing safety criteria, seismicity, improve

availability requirement of in- service inspection, ease of maintenance etc., as

appropriate to the conditions in India.


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DESCRIPTION OF STANDARD INDIAN PHWR:

1) LAYOUT:

The nuclear power stations in India are generally planed as two units

modules, sharing common facilities such as service building, spent fuel storage

bay & other auxiliaries like heavy water upgrading, waste management facilities

etc. . Separate safety related systems & components are however provided for

each unit. Such an arrangement retains independence for safe operation of each

unit & simultaneously permits optimum use of space, finance & construction time.

The lay out for a typical 220MWe station as given in figure 1, shows two reactor

building, active service building including spent fuel bay, safety related electrical

& control buildings and the two turbine buildings. Orienting turbine building

radial to the reactor building provides protection from the effect of turbine

missiles. Other safety related building s &structures are also located as not to fall

in the trajectory of missiles generated from the turbine. The buildings and

structures have also been physically separated on the basis of their seismic

classification.

Sectional views of the reactor building are shown in figure 2 depicting general

layout inside the reactor building.

2) REACTOR:

In concept, the Indian pressurized heavy water reactor is a pressure

tube type reactor using heavy water moderator, heavy water coolant & natural

uranium dioxide fuel. The reactor as shown in the cut away view in figure 3

consists primarily of calandria a horizontal cylindrical vessel. It is penetrated by alarge number of zircaloy pressure tubes (306 for 235MWe reactor), arranged in a


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square lattice. These pressure tubes, also refer as coolant channels, contain the

fuel & hot highpressure heavy water coolant. The pressure tubes are attached to

the alloy steel and fitting assemblies at either end by special role expended joints.

A typical pressure tube assembly is shown figure 4 .Endshields are the integral

parts of the calandria and are provided at each end of the calandria to attenuate the

radiation emerging from the reactor, permitting access to the fuelling machine

vaults when the reactor is shutdown. The end fittings are supported in the end

shield lattice tubes through bearing, which permit their sliding. The calandria is

housed in a concrete vault, which is lined with zinc metallised carbon steel &

filled with chemically treated demineralised light water for shielding purposes.

The end shields are supported in openings vault wall, and form part of the vault

enclosure at these openings. Removable shield plugs fitted in the end fittings

provide axial shielding to individual coolant channels.

3) REACTIVITY CONTROL MECHANISMS:

Due to the use of natural uranium fuel & on- load refueling, the PHWRs do

not need a large excess reactivity. Correspondingly the devices required for

control of reactivity in the core need not have large reactivity worths. Standard

reactors designs are provided with four systems for reactivity control, viz.

1) Regulating rods

2) Shim rods

3) Adjuster rods for xenon override

4) Natural boron addition in the moderator to compensate for the excess

reactivity in a fresh core &for absence of xenon after a long shutdown.

The reactivity control devices are installed in the low- pressure moderator region

& so they are not subjected to potentially severe hydraulic & thermal forces in

the event of postulated accidents. Furthermore, the relatively spacious core lattice


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of PHWR allows sufficient locations to obtain complete separation between

control & protective functions. The regulating systems are thus fully independent

with its own power supplies, instrumentations & triplicated control channels.

Cobalt & stainless steel absorber elements have been utilized in the reactivity

control mechanisms.

For 220MWe standardized design, two diverse, fast acting &

independent shutdown systems have been adopted. This feature provides a high

degree of assurance that plant transients requiring prompt shutdown of the reactor

will be terminated safely. The primary shutdown system consists of 14

mechanical shut off rods of cadmium sandwiched in stainless steel &makes the

reactor sub- critical in less than 2 secs. Fail-safe features like gravity fall &spring

assistance have been incorporated in design if mechanical shut off rods. The

second shutdown system, which is also fast acting, comprises 12 liquid poison

tubes, which are filled with lithium penta borate solution under helium pressure.

The trip signal actuates a combination of fast acting valves and causes poison to

be injected simultaneously in 12 interstitial liquid poison tubes of calandria.

4) FUEL DESIGN:

Fuel assemblies in the reactor are short length (half metre long) fuel bundles.

Twelve of such bundles are located in each fuel channel. The basic fuel material is

in the form of natural uranium dioxide a pellet, sheathed &sealed in thin zircaloy

tubes. Welding them to end plates to form fuel bundles assembles these tubes.

Figure 5 shows the 19- element fuel bundle being used in 220 MWe PHWRs.

5) FUEL HANDLING:


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On power fuelling is a feature of all PHWRs, which have very low excess

reactivity. In this type of reactor, refueling to compensate for fuel depletion & for

over all flux shaping to give optimum power distribution is carried out with the

help of 2 fueling machines, which work in conjunction with each other on the

opposite ends of a channel. One of the machines is used to fuel the channel while

the other one accepts the spent fuel bundles. In addition, the fueling machines

facilitate removal of failed fuel bundles. Each fuelling machine is mounted on a

bridge & column assembly. Various mechanisms provided along tri- directional

movement (X, Y&Z direction) of fueling machine head and make it possible to

align it accurately with respect to channels. Various mechanisms have been

provided which enables clamping of fueling machine head to the end fitting,

opening & closing of the respective seal plugs, shield plugs &perform various

fuelling operations i.e. receiving new fuel in the magazine from fuel transfer

system, sending spent fuel from magazine to shuttle transfer station, from shuttle

transfer station to inspection bay & from inspection bay to spent fuel storage bay.

6) PRIMARY HEAT TRANSPORT (PHT) SYSTEM:

The system, which circulates pressure coolant through the fuel channels to remove

the heat generated in fuel, is referred as Primary Heat Transport System. The

major components of this system are the reactor fuel channels, feeders, two

reactor inlet headers, two reactor outlet headers, four pumps &interconnecting

pipes & valves. The headers steam generators & pumps are located above the

reactor and are arranged in two symmetrical banks at either end of the reactor. The

headers are connected to fuel channels through individual feeder pipes. Figure 6

depicts schematically the relative layout of major equipment in one bank of the

PHT system .the coolant circulation is mentioned at all times during reactor

operation, shutdown & maintenance.


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7) MODERATOR SYSTEM:

The heavy water moderator is circulated through the calandria by aid of a low

temperature & low pressure moderator system. This system circulates the

moderator through two heat exchangers, which remove heat dissipated by high

energy neutrons during the process of moderation. The cooled moderator is

returned to the calandria via. Moderator inlet nozzles. The high chemical purity

and low radioactivity level of the moderator are maintained through moderator

purification system. The purification system consists of stainless steel Ion

Exchange Hoppers, eight numbers in 220MWe contains nuclear grade, mixed Ion

- Exchange resin (80% anion & 20% cation resins) .the purification system is

also utilized for removable of chemical shim, boron to effect start up of reactor

Helium is used as a cover gas over the heavy water in calandria. The

concentration deuterium in this cover- gas is control led by circulating it using a

sealed blower and passing through the recombination containing catalyst Alumina

coated with 0.3% Palladium.

7) FUEL:

The use of natural uranium dioxide fuel with its low content of fissile material

(0.72% U-235) precludes the possibility of a reactivity accident during fuel

handling or storage. Also, in the core there would no significant increase in the

reactivity, in the ever of any mishaps causing redistribution of the fuel by lattice

distortion or otherwise.

The thermal characteristics namely the low thermal conductivity and high specific

heat oh UO2 permit almost all the heat generated in a fast power transient to be

initially absorbed in the fuel. Furthermore, high melting point of UO2 permits

several full power seconds of heat to be safely absorbed above that contained atnormal power.


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Most of the fission products remain bound in the UO2 matrix and may get

released slowly only at temperatures considerably higher than the normal

operating temperatures. Also on the account of the uranium dioxide being

chemically inert to the water coolant medium, the defected fuel releases limited

amount of radioactivity to the primary coolant system.

The use of 12 short length fuel bundles per channels in a PHWR, rather than full

length elements covering the whole length of the core, subdivides the escapable

radioactive facility in PHWR has also the singular advantage of allowing the

defected fuel to be replaced by fresh fuel at any time.

The thin Zircalloy 2/4 cladding used in fuel elements is designed to collapse

under coolant pressure on to the fuel pellets. This feature permits high pellet - clad

gap conductance resulting in lower fuel temperatures & consequently lower

fission gas release from the UO2 matrix into pelletclad gap.


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REACTOR AUXILIARIES

END SHIELD COOLING SYSTEM

There are two End Shields provided at both the ends of calandria performing the

following functions.

(i) Providing supports for calandria tubes and pressure tubes.

(ii) Provides radiation and thermal shielding for fuelling machine vaults so that the

fuelling machine vaults can be accessible during shutdown.

Heat is removed from the end shields to moderator and calandria vault water.

However the bulk of the heat is removed by End shield cooling system.

The basic requirements of the end shield cooling system are:

(i)To maintain calaridria side tube sheet (CSTS) of end shield at an average

temperature of 67deg centigrade.

(ii)To maintain temperature difference between various parts of end shield

within permissible limits.

(iii) To avoid stagnant pockets of coolant, in end shield, which could cause

corrosion problems.

(iv)To avoid overheating and hot spots which could lead to damage of end

shield.

(v)To provide venting of end shield for uniform shielding in accessible and


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S/D accessible areas.

The End Shield Cooling System is a closed loop system Consisting of end shields,

circulating pumps, and heat exchangers. An auxiliary loop exists for the control ofwater chemistry.

There are two end shields where the heat is generated due to radiation and

conduction from other reactors component i.e. End fittings, Feeders, convection

and

radiation across insulation gaps. (Almost 50% of the

heat load is from PHT). A total of 1.4 MW of heat load

exists for each end shield. This heat is removed by

__ circulation of demineralised water through the End

Shields. The End Shields consist of two compartments called front and rear

compartments. DM Water (900 LPM)

enters the front compartment (the compartment facing the calandria) from fiveinlets at the top. Front Compartment is further divided into five separate columns.

DM Water passes through these columns at a velocity of 37.7 cm/sec and flows

into the annulus space between the outer and inner shells of End shield.

CALANDRIA VAULT COOLING SYSTEM

In RAPS calandria vault (the space between the calandria and steel lined structural

wall) is full of demineralised (DM) water. DM water filled calandria vault

provides radiation, biological and thermal shielding, and also acts as heat sink in

case of serious contingency. Filling of calandria vault with DM water eliminated

Argon-41 activity of earlier Indian PHWRs which had air filled calandria vaults

(RAPS 1&2 AND MAPS). This drastically cuts the exposure of public in the

vicinity of Indian Nuclear Power Plants.


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The dimensions of the calandria vault are such that a minimum water thickness of

1.35 meters is ensured between the calandria and concrete vault.

This ensures adequate shielding.

FUNCTIONS OF THE CALANDRIA VAULT COOLING

SYSTEM

i)To remove heat generated in vault water.

ii)To provide thermal shielding and biological shield under all condition.

iii) To maintain uniform temperature in the vault structure below permissible limit

under all condition.

iv) Provide an environment compatible with the material used for components

within vault.

Heat appearing in calandria vault water is removed by a closed loop cooling

system. Water at 42.5deg cen. is distributed through perforated header laid out in

the bottom of the vault and warm water at 46.2deg cen. leaves the vault through

header at the top.

VAPOUR SUPPRESSION SYSTEM

Large pooi of water (2200M3, 2.4m deep) at the basement of the reactor building

is provided to limit peak pressure inside volume Vi during LOCA (Loss of coolant

accident) or MSLB(Main steam line break) by condensing

high enthalpy steam. Volume Vi is connected to the suppression pool via an

annular space between the RB structure wall and inner containment wall.


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The suppression pool is provided with a re circulation system to protect against

corrosion and biological growth.

ANNULUS GAS MONITORING SYSTEM

The annulus gas monitoring system of RAPP 3&4 provides a means of monitoring

the leakage (if any) of heavy water either from PHT or from moderator system

due to failure of coolant tube calandria tube or rolled joints. It is a closed loop

recirculating system which maintains flow of C02 gas through the annulus gap

between coolant tithe and calandria tithe. Apart from leak detection, the annulus

gas also acts as a thermal barrier, separating the hot high pressure coolant tubes

and the comparatively cooler low pressure calandria tubes. By reducing heat

transfer between coolant tube and calandria tube, heat removal requirements from

moderator system are minimized as well as the reduction in loss of heat from PHT

system. In addition, the annulus gas minimizes corrosion and hydrides formation

in the coolant tubes or in the garter spring spacers by providing a dry 02 doped gas

atmosphere in the annulus.

LIQUID POISON INJECTION SYSTEM

For prolonged shutdown of reactor (1) for maintenance jobs or (ii) when reactor

has tripped on reactivity transient which do not permit restart of reactor within

poison override time, LPIS is actuated so that sub criticality margin is maintained

under all conditions. LPIS adds a bulk amount of liquid poison directly to the

moderator to keep the reactor under shutdown state for prolonged duration. This is

an independent process system and is the replacement of (i) ALPAS bulk addition

mode (at NAPP and KAPP) which required moderator circulation and (ii) gravity

addition of boron (GRAB)

The LPIS works on pneumatic pressurization of boron solution by helium. The

system consists of poison tank


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and helium tank. When a command for poison addition is received the pressure

balance valves and siphon break valves close and injection valves open. This

causes the pressurisation of poison tank by helium stored in helium tank. This in

turn causes injection of boron poison directly into the moderator through two

nozzles in calandria at 75%FT level

D2O EVAPORATION AND CLE~AN UP SYSTEM

D20 evaporation and clean up system purifies downgraded heavy water to a level

which is not harmful to heavy water upgrading system by removing all the

impurities. The heavy water collected from various leakages and spills contains a

number of impurities which normally arise fromSurf ace from which D20 is

collected. Corrosion products produced inside the reactor D20 system.

Products resulting from radiolytic process. Organic material from ion exchange

resin dueteration and breakdown.

D20 evaporation and cleanup system is designed to clean the downgraded heavy

water chemically so that it can be fed to upgrading plant. Cleanup system

comprises of oil water separation stage, filtration stage and ion exchange stage.

HEAVY WATER UPGRADING SYSTEM

Heavy water is used as moderator and primary heat transport fluids in PHWRs.

Heavy water is highly hygroscopic. Hence it leaks from the system, it gets

downgraded on exposure to atmosphere. Such leaked heavy water collected from

various points in the reactor is to be upgraded before use in reactor, since the

isotopic purity required for moderator heavy water is as maximum as achievable.

FIRE FIGHTING SYSTEM


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Fire protection system in a nuclear power plant is meant To prevent damage to

various equipment or system due to fire.

To ensure decay heat removal of the reactor. To minimize the release ofradioactivity to environment in the event of a fire.

To provide backup PW cooling to various systems. To ensure personnel spray

supply.

Fire protection system consists of fire fighting water system, carbon dioxide fire

protection system and portable fire protection system.

FIRE WATER SYSTEM

Fire water system comprises of constantly pressurized fire hydrant system and

sprinkler system. Automatic sprinklers have been provided for oil filled

transformers and non-automatic sprinklers are provided for oil systems, cable

vaults and cable tunnels. Hydrant system covers the whole plant for outdoor and

indoor supply of firewater. Water for both hydrant and sprinkler system is

supplied by the firewater pumps from the sumps located in the cooling water

pump house(CWPH).

ACTIVE PROCESS WATER SYSTEM

Active process water system provides direct means of heat transport from

equipment and heat exchangers in the primary heat transport system, moderator

system and reactor auxiliary system to ultimate heat sink during all operational

stages of the plant and accident condition like LOCA. Thus it forms the secondary


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part in the ultimate heat removal system. It is a safety-related system. Reliability

and continuous heat removal is achieved by designing the system for SSE/OBE by

providing redundancy in rotating equipment, Class III power supply to all safety

related electric motor driven equipment and backup supply from fire water system

to meet static component failure This system is potentially active since there is a

possibility of leakage of active primary fluid to this system through various heat

exchangers.

RB VENTILATION SYSTEM

RB is designed as a double containment structure in order to prevent ground level

release during accident conditions. Primary containment houses all equipments

and piping of nuclear systems. Secondary containment envelops the primary

containment with an annular radial gap of 2 meters.

PC is divided into two volumes. Vi containing the systems having high enthalpy

fluids comprising of F/M vaults, pump room, dome region and includes FMSA

when they are in contact with F/M vaults. These areas are not accessible during

normal plant operation. No ventilation is provided for this volume but closed loop

heavy water vapour recovery system is provided to recover D20 that escapes from

high enthalpy systems. The remaining area constitutes volume V2. Volume V2 is

separated from Vi by a leak tight barrier and pressure suppression pool. The

volume Vi is maintained at negative pressure with respect to V2 by maintaining

continuously a small purge to the stack. Volume V2 is normally accessible except

moderator room, FMSA and DN monitoring rooms.

HEAVY WATER VAPOUR RECOVERY SYSTEM

Heavy water vapour arising out of spills/leakages from primary heat transport,

moderator and fuelling machine circuits is recovered from building atmosphere by


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adsorption on molecular sieve beds. Vapour recovery system is an important

feature of the station heavy water management schemes. Following are the criteria

for design and operation of vapour recovery system To effect economy in

reactor operating costs by efficient recovery of heavy water that escapes into the

building atmosphere.

To minimise heavy water loss and tritium loss and tritium release through stack.

To minimise tritium activity levels in various areas of the reactor building.

To keep the volume Vi area under negative pressure with respect to volume V2

area for preventing the spread of activity from volume Vi to volume V2.

CALANDRIA

The calandria is horizontal vessel housed in a rectangular calandria vault. The

calandria is a single walled austenitic stainless steel vessel. The main shell is

stepped down in diameter at each end and site welded to their cylindrical

extensions of the end shields on each side of the reactor.

END SHIELD

The end shields are cylindrical boxes whose extensions are welded to the

calandria side tube sheet at the calandria end and fueling machine side tube sheets

at fueling machine end of the end shield during shop fabrication. The box is

pierced by 306 lattice tubes arranged on 228.6mm square pitch. The space inside

the end shield is divided into two compartments by a 38mm thick baffle plate and


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fueling machine side tube sheet is filled with 10mm dia spherical mild steel balls

and light water in the 57:43 ratio.

CALANDRIA TUBE

The calandria tubes are manufactured from Zircalloy2 strip that is cylindrically

formed and seam welded. The seams are then leveled by rolling. The primary

functions of the calandria tubes in a reactor system are-

1.To separate the relatively cold moderator from hot coolant tubes to minimize

heat losses.

2.To support the horizontal coolant tubes (through garter springs) and prevent the

excessive sag caused by creep. To act as containment vessel for the contents of the

channel in the unlikely but postulated instance of a pressure (coolant) tube rupture

accident.

COOLANT TUBE

Coolant tube is the most important structural component inside the reactor core.

Coolant tubes are manufactured from Zr-Nb. Each end of the coolant tube is

joined to a special type 403SS end fitting. Such 306 Nos. of parallel coolant tubes


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are placed horizontally inside the reactor core at the square lattice distance of 228.

6mm.

END FITTINGS

The end fittings on either end of the reactor identical and connected at the ends.

GARTER SPRING SPACERS

Four numbers of garter spring for each coolant channel and located in the annulus

space between coolant and calandria tubes.

SEAL PLUG

The function of the sealing plug is to close the ends of the coolant assemblies and

prevent the escape of heavy water from the end fittings. During fuel changes it is

necessary to remove these plugs.

SHIELD PLUG

The shield plug which normally resides in the end fitting serves the three

functions of providing Radiation shielding at the ends of the coolant tubes

Means of locating the fuel in the fuel channel and stopping the fuel column from

following the seal plugs when they are withdrawn during fuel changes. The

turbine is of the horizontal tandem compound, reheating, impulse type, running at


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3000 rpm, with special provision for extraction of moisture. The turbine has a

maximum continuous and economic rating of 220 MW, The turbine comprises of

one HP cylinder and two double flow LP

Cylinders thus providing 4 LP flow in parallel. Thetas are five impulse stages in

the HP cylinder and six stages for

each of the LP cylinders. The turbine cylinders and generator is solidly coupled

together in line, with a single thrust bearing on HP shaft between No, 2 bearing

(HP rotor bearing) and the HP~LP. Coupling each rotor is supported in two main

bearing. A solid forged steel rotor is provided in the HP cylinder whilst the LP

rotor have shrunk and keyed on discs, The nozzle plates of the HP cylinder are

welded assemblies incorporating machined nozzle segments, The LP diaphragms

are cast iron with cast-in nozzle division plates. Steams packed labyrinth glands

are provided for each cylinder, Live steam at a pressure of 580 psig and temp

482.60F (saturated) is supplied to the HP cylinder of the turbine through two

separately anchored steam chests each containing a steam strainer a combined

stop and emergency valve and two throttle (or governing) valves, The chests are

connected to the HP cylinder through loop at allow axial movement of this

Cylinder, and ensure that no excessive thrust loads from the piping are transmitted

to the HP cylinder. Extraction steam is taken for feed heating purpose before

stages 4&5 and at the exhaust of the H.P cylinder after expansion is led two

moisture separators in parallel which reduce the moisture content of the steam

before it is reheating is two live steam reheaters, The steam from the reheaters

Having a pressure of 47.5 psig and temperature of 43OoF passes through governor

operated butterfly interceptor valves before entering the two double flow LP

cylinders. An interceptors valve is provided in the line from each reheater to the

LP inlets, The LP cylinder of turbine is of four flow design: each two flow LP has

a central admission belt with outward direction of steam flows. Steam is supplied

to the two flow provision via the separator and reheater in each HP cylinder .the

exhausts from the LP bleeding combines into a condenser, which is maintained at

vacuum 27.5 hg.

Steam is extracted from double flow L1~ cylinder before stages 24 and b for feed

heating and before stage 6 for a moisture extractor, The over all length of the


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Turbine generators 100 and the outside diameter of the last row of blade is 100.

A data logger monitors all turbine and ancillaries parameter.

STEAM CYCLE:

Steam for the turbine through two steam lines or header to the two stem lines or

header to the two combined stop and emergency valves. A10 balance line

connected line connect the header the C.S.E. valves. During normal operation the

C~S.E. valves are fully open to permit steam flow to inlet steam chest and then to

the two governor valves. Governor valve position controls turbine speed and load

and thus are made responsive to the governor valves (two on each bank) are

connected by means of balance lines and the steam passes to the H.P. inlet

nozzles, trough the H.P. cylinder. The to governor valves in each steam chest are

in parallel i.e., there is common inlet and outlet manifold for both of the two

governor valves in a steam chest. Also lines from each steam chest joint and go to

both top and bottom of H.P, cylinder This arrangement in conjunction with the

10 balance line ensures uniform steam take off from each of the 8 boiler and

uniform distribution to each portion of H~P. cylinder, After expansion, the steam

leaves the H~P. cylinder and passes through the separator, reheater, LP emergency

stops valves and interceptor valves, before entering LP cylinders. Also to relief

valves are installed after each of the two re heaters. In the event of governor or

interceptor valve malfunction the relief valve will open and vent the steam toatmosphere preventing over pressurizing the separator or reheaters,

The interceptor valves remain full open normal operation and admit the steam to

LP cylinder from where it is exhausted to the condenser, not all of the steam to theLP cylinders from where it is exhausted to condenser. Not all of the steam


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admitted to turbine by the governor valves is expanded through the turbine and

exhausted to main condenser At different points on the turbine, steam is bled off

or Extracted and passed to feed water heat exchangers. Heating the feed water by

extraction steam has two beneficial results; one is an increase in the heat cycle

efficiency and the minimum permissible inlet feed temp. to boiler is 24OoF, The

RAPS turbine has six extractions feed heaters three including deaerators heaters

Fed from the H.P, cylinder, and are called lob pressure heaters if they are in the

feed line before the boiler feed pumps and are called high pressure if they are in

the feed line after the pumps. Five of the extraction lines have spring closed check

valves. These check valves close on a turbine trip to prevent entrained stream in

the extraction lines and heaters from backing up into the turbine and causing it to

over speed. Entrained steam from the (one, remaining extraction line and heater

was calculated to give only a small increase so check valves were omitted

CONDENSING SYSTEM:

GENERAL.

The circulating water in the condenser condenses the exhaust steam from LP

Turbines. The condensate is recycled through boilers. The air gases are removed

from the condensate by the air ejectors. The condensing system is provided to

supply condensate from the deaerators under all condition of operation. The

maximum flow of condensate to deaerators at 100% turbine load is i

900.000lbs/hr, the design temp. are 91 5oF in the condenser hot well and 245oF at

the deaerators inlet

DESCRIPTION


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Condensate system comprises of main condenser two 100% capacity condensate

extraction pumps and two 21/4% duty emergency pump, 2 moisture Extractor,

gland steam and high level reserve feed water tank with their associate fittings,

pipelines and instrumentation

The condensate extraction pumps take suction from the condenser hot well and

discharge through the moisture extractors, drain cooler and LP heaters, the

condensate flow is controlled by the control valve part of the Condensate front the

condensate pump discharge header flow changes the gland steam condenser and

air ejectors and returns to main condensate line before it enters the moisture

extractors. The flow in this line is controlled by means of regulating control valve

which maintains a fixed differential pressure across the gland steam condenser

having been designed to have the same pressure deferential tar its design flow as

the air ejector. A condensate recirculation line back to the condenser is provided.

This take off is located downstream of the condenser. The condensate pumps also

supply boiler feed pump gland seal water and water f or the turbine spray cooling.

One 21/2% capacity auxiliary condensate

Extraction pump takes water from the condenser hot well and discharge into thesame system as the 100% pumps.

SAFETY DESIGN PRINCIPALS

It has been ensured that systems, components & structures having a bearing on

reactor safety are designed to meet stringent performance & reliability

requirements. These requirements are met by adopting the following design

principles:

a) The quality requirements for design, fabrication, construction & inspection

for these systems are of the high order, commensurate with their importance to

safety.


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b) The safety related equipment inside the containment building is designed

to perform its function even under the elevated pressure & temperature &steam

environment conditions expected in the event of postulated loss of coolant

accidents (LOCA).

c) Physical & functional separation is assured between process systems &

safety systems.

d) Adequate redundancy is provided in systems such that the minimum

safety functions can be performed even in the event of single active components

in the system.

e) To minimize the probability of unsafe failures

f) Provisions are incorporated to ensure that active components in the safety

systems are testable periodically.

g) All the supplies /services (electric, compressed air or water) to these

systems, necessary for the performance of their safety functions are assured &

safety grade sources.

SAFETY & SEISMIC CLASSIFICATION OF SYSTEMS:

SAFETY CALSSIFICATION:

In the design of Indian PHWRs, it is required to grade various systems, equipment

& structures in their importance to safety & reliability. The safety gradation

consists of four different safety classes depending upon the nature of safety

functions to be performed by the various items of the plant.


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SAFETY CLASS I: It is the highest safety class & includes equipment &

structures needed to accomplish safety functions necessary to prevent release of

substantial core fission product inventory. This includes reactor shutdown systems

& primary heat transport system.

SAFETY CLASS II: Includes equipment, which performs those safety

functions, which become necessary to mitigate the consequences of an accident

involving release of substantial core fission product inventory from fuel. This

class also includes those items, which are required to prevent escalation of

anticipated operational occurrences to accident conditions. Boiler feed water &

steam system, emergency core cooling system, reactivity control provisions &

reactor containment building are included in this class.

SAFETY CLASS III:

Includes systems that perform functions, which are needed to support the safety

functions of safety class II & I. Also, it includes systems & functions required to

control the release of radioactivity from sources located outside the reactor

building. Process water-cooling system include induced draft cooling towers, air

supply system, shield cooling system primary coolant purification ion exchange

columns & filters etc. are included in this category.

SAFETY CLASS IV:

Includes those items & systems, which do not fall within the above classes but are

required to limit the discharge of radioactive material & airborne radioactivity

below the prescribed limits .D2O upgrading, waste management, dueteration

&service building ventilation systems are classified as class IV safety systems.


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40


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1.0 DESIGN DESCRIPTION:

COIS is a data acquisition and display equipment for providing the operator with process

alarm messages, status, trend curves, history displays and printouts of groups of process

variables etc.

. A three-tier system design consisting of Display Stations, Data Acquisition Computers

and I/O subsystem has been adopted to improve the reliability and availability. This

makes the different subsystems to be hardware independent on each other. A high speed

Ethernet LAN (Local Area Network) is used for communication between the subsystems.

10 of the Display Stations are Utility CRTs (UCRTs) and the remaining 2 are Alarm

CRTs (ACRTs). They are Intel 80386 microprocessor based systems doing most of the

user interface job. These systems are having high resolution (Super VGA-1024 x 768

pixels) 19 monitors, which give a good pictorial representation ofthe data.

Data Acquisition Computers are based on Intel Pentium which UNIX SVR 4.2 as the

Operating system. Both the DACs work in dual redundant hot standby mode. They

mainly acquire the data from I/O subsystems and other Computer Systems like PLC,

DPHS, RADAS etc. and pass the required data to display stations. They do the logging

of the history data and take care of the printer tasks. They also do the network

management of both the LANs. They also direct the I/O systems to govern the field

outputs as required.

I/O subsystems are Motorola 68020 CPU based systems. Each I/O subsystem has 2 CPUs

working in dual redundant mode. They mainly do the scanning and alarm checking of

the field inputs connected to them and pass on the data to DACs. They also change the

field outputs as per the directive from DACs.

The network topology is designed in such a way that a single break anywhere in the

network (broken cable or failed n/w equipment) will not result in a collapse of the total

system, but will allow the system to continue to work at a degraded level. The significant

aspects of the designed network topology are:


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a). Thicknet cable is used as it is much more rugged than the thinnet cable.

b). Transceivers that are used to connect different nodes to the network are piercingtype tap boxes, which facilitate the connection without a cut in the cable.

c). A significant component in the topology is Repeater. Repeater is an active

component that can be used to connect different cables of networks. It isolates the

remaining network from a fault in any of the other cables. This has given rise to a fault

tolerant network. The network is divided into 4 parts each connecting of the system.

The various failures considered their effect is described below:

i). If any transceiver or the cable connecting the transceiver to the node fails, only

that node will fail & reset of the system will continue to work as usual. If the connection

with the Master DAC fails, the hot standby DAC will take over and the system will not

be affected.

ii). If any one of the cables of LAN1 - a, b, c or d fails than of inputs/outputs (of

the nodes connected to that part of the network) will be lost & the system will continue to

work with 75% of inputs/outputs data. If more than one cable fails, COIS still will work

with the reduced capacity accordingly.

iii). If any one of the cables of LAN-2a, b, c or d fails, then of display stations

will not be available. Display stations are connected in these four cables in such a way

that CRTs on adjacent Main Control Room panels are connected to different cables and

will not fail simultaneously. ACRTs are connected to different repeaters and hence will

not fail simultaneously. If more than one cables fail, COIS will still work with the

corresponding reduction in display stations.


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1.1 Inputs/Outputs

There are various types of plant inputs to the COIS viz. analog inputs and digital inputs.

Each plant input is also referred to as point the COIS also provides voltage free relaycontacts as outputs.

Analog inputs

There are 1256 analog inputs to the COIS. These include about 10% spare points. The

approximate distribution into different categories is as follows: -

RTDS 392 points

Thermocouples 16 points

Volts/Current Inputs 824 points

Thermistor 8 points

Exact details of description, input range, alarm priority, process range and processing

required etc. for each analog input are available in the COIS analog input..

For the current inputs the terminating resistors (of value as specified in analog input list)

are a part of the COIS. Among the 824 voltage or current inputs, any number may be


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44

voltage input. Thus all of these 824 inputs can be arranged to take a current or voltage

input. Input impedance of voltage inputs is greater than 1 Ohm. Linearisation and lead

resistance compensation wherever necessary, e.g. for RTD and thermocouple inputs, will

be performed by the COIS. All RTD inputs will use 3 wire RTDs in the field. Provision

will be use 3 wire RTDs in the field. Provision will be made to terminate a 3 wire RTD ateach RTD input point.

Cold junction compensation for thermocouple will also be provided by the COIS. These

analog inputs are numbered in the range of 000 to 1299.

1.1.1 Digital (contact) Inputs

There are 1136 digital inputs of contact type. There are two types of contact inputs as

follows:

1.1.1.1

Voltage free contact inputs: There are 736 voltage free contact inputs. These

contacts represent alarm or status inputs.

1.1.1.2 Shared contact inputs: There are 400 field contact inputs which are shared

between the window Annunication system (WAS) and the COIS. These contacts

represent alarm events.

1.1.2

Digital (Voltage level) Inputs

There are 656 voltage level inputs representing the status of valves (open or closed).

State Voltage

Level 0 Between 0 volts and 2 volts


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45

Level 1 Between 40 volts and 48 volts.

The input impedance presented by COIS to these inputs will not be less than 10K ohm.

1.1.3

Input Data from other computer systems

COIS receives data from other computers viz. Digital Recording System (DRS),

Radiation Data Acquisition System (RADAS), Electrical DAS and PLCs via LAN

thorough gateways. In the COIS, these input parameters are numbered as follows:

(1) Radiation Data Acquisition System:

Analog points: 3601 to 3799

Digital points: 3001 to 3999

(2) Electrical DAS:

Analog & contact points: 7001 to 9499

(3) PLCs Digital points representating: 1301 to 1999

Status of hand switch position & 5901 to 5999

(4)Digital Recording System (DRS)

a) Normal/Disturbance Analog inputs : 2801 to 2899

b) Visicorder Function Analog inputs : 2901 to 2999

c) Contact inputs of ESR function : 9501 to 3499

d) Dual Process Hot Standby : 9501 to 9699 Analog inputs

(5)Other Computer Systems

a) Analog inputs : 9701 to 9899

b) Contact/digital inputs : 9901 to 9999


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Note: There are no physical inputs corresponding to these points. Values of these points

are provided to the COIS periodically by the above systems. For all displays and logging

functions except alarm functions, these points are treated as the field COIS inputs.

1.1.4 The COIS will also provide 224 outputs of voltage free relay contacts. Ten or

these contacts are used for Fuel Failure Monitoring function described in Sec. testing

function described in section 5.11. The remaining contacts will be used for miscellaneous

purposes like giving time synchronizing pulses to other computer based systems,

annunciation of the failure of the COIS, indication of which computer system is faulty

etc.

1.1.5

Ethernet

The COIS will provide Ethernet LAN connectivity for connection to other computer

systems.

1.1.6 The COIS will also provide, on operators demand, processed data outputs called

calculated analog variables, for e.g. selected channel differential temperatures and DNM

detector outputs etc. These variables will be numbered in the range of 6001 to 6699.

1.2 Accuracy, Noise Rejection, Contact Debounce and Isolation.

1.2.1 For all digital and analog inputs, a very high isolation between the transducer

circuit and the COIS is provided to avoid problems in the transducer circuit due to ground

faults etc.


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1.2.2 Overall accuracy of analog data acquisition for any point will be 0.25% of span or

better. This accuracy will be maintained even in the presence of common mode noise

(Max.) + 15V d.c./50Hz. on the input. A low pass filter will be provided on each analog

input to suppress normal mode (predominantly) 50Hz. noise. Protection will be provided

against following conditions for different categories of analog input.240V AC (Max.),

50Hz common mode voltage on any thermocouple inputs.

For RTD inputs: Depending on RTD bridge excitation network

Or + 50V d.c./a.c. (Whichever is more) common mode or normal

mode voltage.+ 250V d.c./a.c. common mode or + 50V dc/ac

normal mode voltage on any other type of analog inputs.

1.2.3 Processing of potential free contact type inputs will not be affected even under a

max. common mode voltage of + 15V d.c./50Hz a.c. on any input. Beyond 15V,

protection is provided for a max. + 250V d.c. /a.c. common mode voltage. For providing

the above common mode voltage capabilities, opto-isolators are used for digital inputs.

1.2.4 In case of digital (contact) inputs and digital (voltage level) inputs input status

changes lasting less than 50 milliseconds will be ignored by the system.

Note: Sampling interval for all analog & contact inputs will be adjustable to any

of the following: 5 sec. 10 sec., 30 sec., with the normal interval specified in the above

table. This adjustable could be on an individual basis or on group basis (Group = type of

input).

For all analog inputs, five samples will be taken within the sampling intervaland the average of these five samples will be taken as the value for that sampling interval.

There will also be a provision for averaging over last five sampling intervals for

selected number of analog inputs (max. 100Nos.). These average values will be used in

all displays and printouts.


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1.3

Alarm Function

The computer system will check some of the analog inputs and almost all of 736 digital

alarm inputs in the 656 Digital (voltage) inputs for alarm events. Many analog points are

only for periodic logging and BG display etc. and are not checked for alarm at all. And

some points have to be checked for alarm at all. And some points have to be checked for

only low alarm limit or only high alarm limit i.e. both alarms are not required. Some

COIS points are inhibited from reporting to ACRTs as alarms, i.e. these points are not

displayed on the ACRTs when the status of these points changes. But this does not

prevent them from displaying their status in BGs, tabular trends etc. The remaining point

will have both low and high alarm limits. Some of contact (digital) inputs are for status

monitoring only and will not be checked for alarm. The remaining points will be checked

for alarms. The frequency of alarm checking will be same as that or the point for data

acquisition. The alarm events are defined as follows:

1.

An analog input going above a high limit (HL) or falling below a low limit (LL)

or a digital input sent to alarm state since previous scan is referred to as an Alarm

occurrence.

2. Analog input returning between its high and low point or a digital input going to

normal state since previous scan is referred to as return to normal occurrence. This is

also referred to as Normal occurrence is short.

As for as ACRT display or output on alarm printer are concerned, analog points will be

limited to only low, High and Bad states. BL (Bad Low) than lower end of span

Lost. Such additional alarms will store in the computer memory and arranged as CRT

concealed alarm pages for display purposes.

Capacity for such 100 additional alarms is provided. A suitable audiovisual indication for

the alarm in the concealed pages is provided. Operator will be able to call up for display


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any of the ACRT concealed alarm pages on any of the two ACRTs or on both ACRTs

will be different and independent. Provision will be made for scrolling up or down (one

line at a time) of ACRT display. Latest alarm/ return to normal message line will also

be displayed on the last line of all the ACRT pages. Provision will be made so that the

operator can retain on the screen the most important/of immediate interest/relevantalarms only on the screen and put the rest of them in concealed alarm pages. Provision

will also be made to list the alarms on any UCRT for a selected USI or a range of USIs

keyed in by the operator. This is called alarm display management. Each input point

will be given a priority number of 1 or 2. Inputs with priority of 2. It will be possible for

the operator to call up the summary of existing alarms on any UCRT. It will also be

possible to call this summary as total or only of alarms with a priority 1 or only of alarms

with a priority 2. Provision will be made to list all alarms for a selected USI or range of

USIs keyed in by the operator.

Operator will be able to tell the COIS any analog/digital inputs which are to be ignored

(i.e. as if those do not exist) for alarm function. Such ignored inputs will be resumed

automatically in 30 minutes or whenever desired by the operator, whichever is earlier.

Such operator commands will get immediately logged on Alarm Printer.

The system will maintain a list of such alarm-disabled points. It will be possible to

add/delete points in this disabled list.

The COIS will have the following two codes of alarm processing.

1) Mode 1: Under this mode repetitive alarms are suppressed.

2) Mode 2: Under this mode, repetitive alarms are also reported (without any

suppression) lost. Such additional alarms will be stored in the computer memory and

arranged as CRT concealed alarm pages for display purposes.

Capacity for such 100 additional alarms (i.e. approx. 5 concealed alarm pages) is

provided. A suitable audiovisual indication for the alarm in the concealed pages is

provided. Operator will be able to call up for display any of the ACRTs or on both

ACRTs i.e. the pages selection keys for both the ACRTs will be different and


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50

independent. Provision will be made for scrolling up or down (one line at a time) of

ACRT display. Latest alarm/ return to normal message line will also be displayed on

the last line of all the ACRT pages. Provision will be made so that the operator can retain

on the screen the most important/of immediate interest/relevant alarms only on the screen

and put the rest of them in concealed alarm pages. Provision will also be made to list thealarms on any UCRT for a selected USI or a range of USIs keyed in by the operator. This

is called alarm display management. Each input point will be given a priority number

of 1 or 2. Inputs with priority 1 being more important than those with a priority of 2. It

will be existing alarms on any UCRT. It will also be possible to call this summary as total

or only of alarms with a priority 1 or only of alarms with a priority 2. Provision will be

made to list all alarms for a selected USI or range of USIs keyed in by the operator.

Operator will be able to tell the COIS any analog/digital inputs which are to be ignored(i.e. as if those do not exist) for alarm function. Such ignored inputs will be resumed

automatically in 30 minutes or whenever desired by the operator, whichever is earlier.

Such operator commands will get immediately logged on Alarm Printer.

The system will maintain a list of such alarm-disabled points. It will be possible to

add/delete points in the disabled list.

i) Mode 1: Under this mode repetitive alarms are suppressed.

ii)

Mode 2: Under this mode, repetitive alarms are also reported (without any

suppression). The operator through a password can select alarm-processing mode 1 or 2.

1) Alarm processing under mode 1:

Under some abnormal field conditions, some of the points (analog/contact) may oscillate

between alarm and normal states and hence any cause large number of alarm/normal

messages on the alarm printer & ACRTs. Hence, it is required that not more than six

message (status changes) are generated by any point in any quarter of an hour. For this

purpose, the COIS will set the status change count of each to zero, every quarter of an


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hour. Any point which changes status (from normal to alarm/bad or alarm/bad to normal

etc.) 6 times in any quarter of the hour, will be automatically disabled from alarm

scanning for the remaining part of the quarter hour. But the COIS will freeze the status

only with alarm/bad status i.e. if 6th

message is normal message, the COIS will disable

alarm scanning of the point after 7th

message (i.e. Alarm message) is reported. Thisperiod for checking max. NO. Of alarm generation will be programmable between of

an hour to a selected period will also be adjustable between 4 and 10.

Alarm processing under mode 2 :

All alarm messages are reported without any suppression.

1.3.1 There are about 400 Nos. digital inputs (shared input contacts) which are scanned

only for the logging of their status changes on the alarm printer and mag. Tape

cartridge/disk cartridge (i.e. CRT display and audio ann. Is not required for these). (Note:

These are the window annunciator points numbered in the range of 4001 to 4999). These

are scanned every one second.

1.3.2

Latest Alarm Message Display Function on UCRTs

The latest Alarm/Normal message displayed on the ACRTs will also be displayed on

the bottom-most line of all the UCRTs also. No flashing of the alarm message or any

audio is required on the UCRTs. However, successive alarm message will be displayed in

alternate red and pink colours on UCRTs (e.g. first alarm message in red colour, second


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in pink colour and third in red again and so on). Normal message will always be shown in

green colour. Any alarm/normal message will be continuously displayed until another

alarm/normal message is generated to replace the previous one. Operator will provide a

facility to switch off this alarm/normal message display on any UCRT whenever

required.

1.3.3

Valve Status Monitoring

There are 656 Nos. voltage level inputs representing the valve status (open or closed).

These inputs are scanned once every 1-second for displaying the actual status of the

valves in the Mimics and for logging their status changes on printer used for alarm

logging. Status changes are recorded on magnetic tape also.

Each valve to be monitored for its status on COIS will have one or two inputs connected

to COIS. These are voltage level inputs with two levels, i.e. at 0 volt for 48 volts dc.

Voltage level inputs are normally taken across the indicating lamps corresponding to the

valve. When the indicating lamp is ON indicating valve Fully open or fully closed,

the input to COIS is +48V DC, otherwise it is zero. If there are tow inputs corresponding

to a valve, the COIS will sense both the inputs and derive the status as follows:

Fully open input Fully Closed Input Status

48 V 0 V Fully open

0 V 48 V Fully closed

0 V 0 V In intermediat

position

48 V 48 V INST. FAILURE


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The COIS will show the actual status in the Mimics appropriately. It will also log the

change of status on the printer accordingly.

For valves having single input to COIS, the status will be either Fully open or Not

Fully open (Not Fully Closed).

If there is a change in status the new status will be logged on to the printer. It may be

noted that the Intermediate status and Instrument failure status are taken as new

status only if it has remained so for two successive scans. The Instrument failure status

is treated as an alarm and would be annunciated on the ACRT.

1.4

Interface to Various other Computer Based Systems

The COIS will also provide Ethernet LAN interface for connecting the following

computer systems. The COIS will receive data from them as required as per the approved

protocol. The data will be available for all the functions described in this design manual

except for alarm function:

a) RADAS

b) Electrical SCADA (EDAS).

c) PLC

d) DRS

e) DPHS

f) CTM

g) PDCS etc.


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1.5 General Features:

1. The COIS will be user friendly and the operator will be able to get the desired

information in desired formats on the various UCRTs in an interactive manner. Menudriven CRT based dialogue with the system will be designed. Various menus/indexes/lists

of UCRTs functions, BGs, History groups and graphic trend groups etc. will be displayed

on the UCRT on operators demand. Help facility will be available at all phases of

dialogue. Data retrieval procedures will be quick and easy. No UCRT will have a blank

screen at any time. If no display is demanded, it will be showing the main menu, so that

operator can quickly select the display of his choice.

2. Process data base will contain flags for each Analog input to indicate whether any Low

Alarm Limit/High Alarm Limit is applicable or not. Software for Alarm processing and

various displays etc. will not call for entry of artificial low/high alarm limit e.g. lower than

lower end of span/higher than higher end of span etc. as this causes confusion and

inconvenience to the operator while studying the printouts/displays.

4. Microprocessor based and standalone type I/O subsystems are used. Analog inputs

cards are designed for automatic calibration at regular intervals under software control

using precision reference sources for 25% and 75% scale. Hence, software offset correction

is provided for any drift due too temperature or time.

5. A facility for enabling/disabling low alarms in bulk for certain points with a single

command will be provided. These low alarms mostly correspond to the failure of the

sensors.


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1.6

Resolution of Values in Numeric/Plot/Bar Graph Form

Resolution of Values (i.e. data such as current value of a process variable, alarm point, etc.)

will in various cases be as follows or better:

Sr. No. Type of output Resolution

1 Numeric form 0.1%

2 Graphic Display/Mimic/Bar Graph/Plot 0.2% or better

Numerical resolution will be limited to 0.1 to 0.2% of full span. The operator will not

require a better resolution than this and it will waste the useful area of the screen. Hence

numerical displays like 21.3245 for a process range of 0-100C will be avoided and an

approximated value 21.3 will be displayed.

1.7

Response Time

Expected response times are described as follows:

1.

Initial Display Lag: It is defined as the Maximum Time taken for the firstcomplete display (static + dynamic parts) to appear on the display after the operators

common will not be more than 2 seconds.


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2. Time Stamp Lag: It is defined as the maximum time difference between the

real time of a field event occurrence and the time stamp (Time stamp will be done as

soon as the scanning is done) will in principle be same as the sampling interval.

3. Display Lag: It is defined as the maximum time difference between time

stamping of an event and it being displayed will not be more than 1 second.

4. Sampling interval: It is defined as the maximum time between the two

consecutive scanning of the inputs

5. Print Lag: It is defined as maximum time difference between the

commencement of the demanded printout and the operators commands and will not be

more than 10 seconds.

1.8

Data Storage/Retrieval and Off-line Computer System

1.8.1 Data Storage

The following on-line data will be recorded on the magnetic disk for last 32 hours:

a) History data

b) Static data base

c) Changes done in static data base

d) Five sets of DNM data & ECCS Test Data

e) Alarm logging

f) Snapshot of current values of all the points at every shift and as demanded by the

operator.

g) Data and time of recording the data.

This data for the last 24 hours, on default, will be dumped on to the magnetic tape one in

a day at a fixed time, which will be adjustable on system starting time.

Typically, one tape will be used for a day and previous one months data will be stored

(i.e. 31 tapes will be available). Before dumping the data, system will check, if the tape


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on the drive is of that days tape. If not, it will ask the operator to insert new tape for

dumping the data.

1.8.2

Data Retrieval

Provision will be made to retrieve the data from the disk (current 24 hours) or from any

previously recorded magnetic tape and store it on a PC (MS-DOS) compatible floppy.

Facility will be provided to select any type of data and in any range (time, usi, etc.).

1.8.3

Off-line computer system

Off-line computer system will comprise of a standard PC-AT and a printer. It will be

possible to read the data recorded on floppies by the on-line system .The software would

also include standard package like DBASE IV.

1.9

Input Power Supply to the Equipment and Effect of Power

Failure.

Two independent sources of single phase, A.C. power supply of the following

specifications will be available in the station.

Voltage

RMS Value : 240 Volts + 10%

Steady state variation : + 10%

Transient variation : 0% (for 200m secs)

Frequency

Frequency : 50 Hz

Steady state variations : + 1%

Transient variation : + 5%

During transient variations (upto 200m sec) the COIS will continue to operate without

malfunctioning.

An input a.c. power interruption (total outage) lasting 30 milliseconds or less will not

affect the working of the COIS in anyway.


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For all 240V AC loads, both sources of main supply will be connected through

contractors such that failure of any main power supply will not affect operation of any

subsystem. Wherever duplicated D.C. power supplies are used, separate a.c. main source

will be given to the two D.C. power supplies will be connected to the load through

isolation diodes.

1.9.1

Seismic specification

1. The I/O subsystem equipment will operate satisfactorily during and after the

vibration tests at the following peak accelerations when subjected to a sinusoidal

acceleration for 30 seconds at each frequency in the given range.

Peak acceleration I the horizontal axes and vertical axis: 3.5g from 1 Hz to 33 Hz.

1.10 Master Clock Time

Real time clock of the COIS Unit-1 will be used as the master clock for synchronizing

time of various computer systems of the plant. The COIS will provide a potential free

change over contact to each of the computer based systems for time synchronization with

0.5 second status change at 10.00 hours every day (Normal status will be resumed at

10:00:00 hours). The contact status change will be sensed by each of the computer based

systems and the time will be set to 10:00:00 hrs.

1.11

Reliability and Availability

The meantime between failures (MTBF) of the system excluding the printers, plotters andCRTs will be 4000 hours or more with an availability of 99.9% or better. MTBF and

availability figures for the printers/plotters and CRTs will be 2000 hours and 99%

respectively. The meantime to detect a fault (MTTD) plus the mean time to detect a fault

(MTTD) plus the mean time to repair (MTTR) will not exceed 1 hour. In order to achieve


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the high availability figures, a master and hot standby computer redundancy is employed.

In case of a computer going faulty, its load will be automatically switched over to the

other computer. Displays/printouts active before the failure of computer, become active

automatically without operators intervention after switching over pertaining to History

will not be lost due to such switchover. To keep the MTTD+MTTR under one hour, hotrepairs concept is used.

CONCLUSION

As an undergraduate of the Rajasthan University I would like to say that this

training program is an excellent opportunity for us to get to the ground leveland experience the things that we would have never gained through going

straight into a job. I am grateful to the Rajasthan University and APEX forgiving me this wonderful opportunity.

The main objective of the industrial training is to provide an opportunity toundergraduates to identify, observe and practice how engineering is

applicable in the real industry. It is not only to get experience on technicalpractices but also to observe management practices and to interact with

fellow workers. It is easy to work with sophisticated machines, but not withpeople. The only chance that an undergraduate has to have this experience is

the industrial training period.

I feel I got the maximum out of that experience. Also I learnt the way of

work in an organization, the importance of being punctual, the importance ofmaximum commitment, and the importance of team spirit. I have learnt

many things in this 45 days training session. In my opinion, I have gainedlots of knowledge and experience needed to be successful in a great


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engineering challenge, as in my opinion, Engineering is after all a

Challenge, and not a Job.

References

http://en.wikipedia.org/wiki/Rajasthan_Atomic_Power_Station

Power Point Presentation Slides Prepared by me


https://www.google.co.in
http://en.wikipedia.org/wiki/Rajasthan_Atomic_Power_Stationhttp://en.wikipedia.org/wiki/Rajasthan_Atomic_Power_Stationhttp://en.wikipedia.org/wiki/Rajasthan_Atomic_Power_Station


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Documents

j!t Final Report